[Federal Register Volume 62, Number 78 (Wednesday, April 23, 1997)]
[Notices]
[Pages 19820-19823]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 97-10522]
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NUCLEAR REGULATORY COMMISSION
[Docket Nos. 50-255, 50-266/301, 50-313/368, 72-5, 72-7, 72-13]
Consumers Power Company, Palisades Nuclear Plant, Wisconsin
Electric Power Company (Point Beach Nuclear Plant, Units 1 and 2),
Entergy Operations, Inc. (Arkansas Nuclear one, Units 1 and 2),
Issuance of Director's Decision Under 10 CFR 2.206
Notice is hereby given that the Director, Office of Nuclear Reactor
Regulation, has issued a Director's Decision concerning a Petition
dated September 30, 1996, filed by Citizens' Utility Board (Petitioner)
under Section 2.206 of Title 10 of the Code of Federal Regulations (10
CFR 2.206). The Petition requested that the NRC (1) Require Wisconsin
Electric Power Company to retain 24 empty and available spaces in the
Point Beach Nuclear Plant spent fuel pool to accommodate retrieval of
spent fuel from a VSC-24 cask, and (2) prohibit loading of VSC-24 casks
until the Certificate of Compliance, the Safety Analysis Report, and
the Safety Evaluation Report are amended to contain operating controls
and limits to prevent hazardous conditions.
The Director of the Office of Nuclear Reactor Regulation has
determined that the Petition should be denied for the reasons stated in
the ``Director's Decision Under 10 CFR 2.206'' (DD-97-09), the complete
text of which follows this notice. The decision and documents cited in
the decision are available for public inspection and copying in the
Commission's Public Document Room, the Gelman Building, 2120 L Street,
NW., Washington, DC.
A copy of this decision has been filed with the Secretary of the
Commission for the Commission's review in accordance with 10 CFR
2.206(c). As provided therein, this decision will become the final
action of the Commission 25 days after issuance unless the Commission,
on its own motion, institutes review of the decision within that time.
Dated at Rockville, Maryland, this 17th day of April 1997.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
Director's Decision Under 10 CFR 2.206
I. Introduction
On September 30, 1996, Citizens' Utility Board filed a Petition
pursuant to Section 2.206 of Title 10 of the Code of Federal
Regulations (10 CFR 2.206) requesting that the U.S. Nuclear Regulatory
Commission (NRC) take the following actions:
1. Order Wisconsin Electric Power Company (WEPCO) to retain 24
empty and available spaces in the Point Beach Nuclear Plant spent
fuel pool to provide the capability to permit retrieval of spent
fuel from a VSC-24 cask in the event of an accident requiring
removal of spent fuel from the cask or in the event that conditions
of the certificate of compliance (COC) for the VSC-24 require
removal of spent fuel from the cask, until such time that WEPCO has
other options available to it to remove spent fuel from a cask in
the event conditions warrant it; and
2. Order users of the VSC-24 cask not to load VSC-24 casks until
the COC, safety analysis report (SAR), and safety evaluation report
(SER) are amended to contain operating controls and limits that
prevent hazardous conditions, including but not limited to the
generation of explosive gases, due to VSC-24 material reactions with
environments encountered during loading, storage, and unloading of
the VSC-24 cask. The SAR and SER must be amended such that each
operating control and limit is clearly documented and justified in
the technical review sections of the SAR and associated SER as
necessary and sufficient for safe cask operation.
The Petition has been referred to me pursuant to 10 CFR 2.206. The
NRC letters dated October 11 and December 10, 1996, to Mr. Dennis Dums,
on behalf of the Petitioner, acknowledged receipt of the Petition and
provided the NRC staff's determination that the Petition did not
require immediate action by the NRC. Notice of receipt was published in
the Federal Register on December 16, 1996 (61 FR 66063).
[[Page 19821]]
On the basis of the NRC staff's evaluation of the issues and for
the reasons given below, the Petitioner's requests are denied.
II. Background
The Petitioner's first request is for the NRC to order WEPCO to
maintain sufficient empty space in the spent fuel pool at Point Beach
to accommodate the unloading of a VSC-24 spent fuel storage cask. NRC
regulations include a requirement that an independent spent fuel
storage installation (ISFSI) be designed to provide for the ready
retrieval of spent fuel or high-level radioactive waste for further
processing or disposal. This requirement is applicable to ISFSIs so
that the stored spent fuel can be retrieved for transport to either a
monitored retrievable storage installation (MRS) or a high-level waste
repository whenever it becomes available. This regulation, 10 CFR
72.122(l), provides as follows:
(1) Retrievability. Storage systems must be designed to allow
ready retrieval of spent fuel or high-level radioactive waste for
further processing or disposal.
In addition to the regulatory requirements in Section 72.122(l)
pertaining to retrieval of the fuel assemblies for further processing
or disposal, there are certain events or conditions that could warrant
removing a VSC-24 cask from an ISFSI and returning the multi-assembly
sealed basket (MSB) to the spent fuel pool and unloading the stored
fuel assemblies. The COC requires a VSC-24 cask to be returned to the
spent fuel pool in response to those design basis events or conditions
that may challenge the integrity of the storage cask or the cladding of
the spent fuel assemblies.1
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\1\ The following sections of the COC include requirements for
returning a VSC-24 cask to the spent fuel pool and/or unloading the
cask:
Section 1.2.3, ``Maximum Permissible Air Outlet Temperature'';
Section 1.2.10, ``Time Limit for Draining the MSB'';
Section 1.2.15, ``Handling Height''; and
Section 1.3.4, ``Thermal Performance.''
Each section is discussed later in this decision.
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Petitioner's second request is for an NRC order to WEPCO and other
users of VSC-24 casks not to load additional casks until the COC, the
SAR, and the SER are amended to contain operating controls and limits
to prevent hazardous conditions. On May 28, 1996, a hydrogen gas
ignition occurred during the welding of the shield lid after spent fuel
had been loaded into a VSC-24 cask at the Point Beach Nuclear Plant.
The hydrogen was formed by a chemical reaction between a zinc-based
coating (Carbo Zinc 11) and the borated water in the spent fuel pool.
Following the event, the NRC issued confirmatory action letters (CALs)
to those licensees using or planning to use VSC-24 casks for the
storage of spent nuclear fuel (i.e., licensees for Point Beach,
Palisades, and Arkansas Nuclear One). The CALs documented the
licensees' commitments not to load or unload a VSC-24 cask without
resolution of material compatibility issues identified in NRC Bulletin
96-04, ``Chemical, Galvanic, or Other Reactions in Spent Fuel Storage
and Transportation Casks,'' dated July 5, 1996, and subsequent
confirmation of corrective actions by the NRC. The staff has
acknowledged that the event demonstrated that the SAR and related NRC
review, as documented in the SER, did not adequately address the use of
a zinc-based coating and its reaction with the acidic water in spent
fuel pools.
The licensees using VSC-24 casks submitted to the NRC information
on operating controls and limits to prevent hazardous conditions
implemented in response to NRC Bulletin 96-04 and subsequent staff
inquiries. The submittals from the licensees included evaluations of
possible material interactions and provided descriptions of how
procedures were revised. The revisions include controls for the
environments that the casks encounter during use, requirements for
inspections and environmental sampling, and additional precautions for
various cask operations. The NRC staff has evaluated these responses
for Arkansas Nuclear One (ANO) and Point Beach and, as documented in
the safety evaluations dated December 3, 1996, and April 8, 1997,
determined that the operating controls and limits proposed by these
licensees are acceptable and satisfy regulatory requirements. By a
separate letter also dated December 3, 1996, the staff informed the
licensee for ANO that its corrective actions had been verified by
inspections performed by the NRC staff. Shortly thereafter, the
licensee initiated cask loading activities.2 The NRC will
perform inspections in the near future in order to verify corrective
actions implemented at Point Beach. The review of responses to the
bulletin related to Palisades is ongoing. Cask operations at Point
Beach and Palisades continue to be limited by the licensees'
commitments described in CALs.
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\2\ The NRC staff is looking into reports from licensees on the
need to perform weld repairs during the welding of the shield lid
into the MSBs of several VSC-24 casks. This potential problem is not
related to the requested actions or supporting information cited in
the Petition. The NRC staff determined that the issuance of this
Director's Decision should not be delayed pending resolution of
potential problems associated with the weld repairs because the weld
repairs are not related to concerns presented in the Petition and
the welding issue is being addressed by ongoing NRC activities. The
Petitioner was informed of the welding issue and the NRC staff's
decision to not include the issue in the staff's evaluation of the
Petition.
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III. Discussion
As noted, the Petition requests two actions be taken by the NRC.
They are addressed below.
Item 1: Order WEPCO To Retain 24 Spaces in the Point Beach Spent Fuel
Pool
The first requested action calls for the NRC to issue an order to
WEPCO to retain 24 empty and available spaces in the Point Beach spent
fuel pool to provide the capability to unload a VSC-24 dry storage
cask. The two basic reasons to return a cask to the spent fuel pool
would be either to (1) Retrieve the fuel assemblies for further
processing or disposal pursuant to 10 CFR 72.122(l) or (2) respond to
an event or condition that has potentially degraded the cask or spent
fuel in regard to the requirements established in the COC.
As previously discussed, 10 CFR 72.122(l) sets forth requirements
pertaining to retrieval of the fuel for further processing or disposal;
however, it provides no basis for the NRC to require a licensee to
maintain a specified reserve capacity in the spent fuel pool. Licensees
will have considerable opportunity to plan and schedule the activities
associated with retrieving fuel assemblies from existing storage casks
for transfer to other casks for further processing or disposal. This
ability to control the activity includes either ensuring that existing
spent fuel pool facilities will support the transfer or developing
alternate approaches. Alternate approaches could involve, for example,
making room in spent fuel pools by use of other storage or
transportation casks, expanding the wet storage capacity by making
changes to the spent fuel pool or other parts of the reactor facility,
or development of a system for direct cask-to-cask transfer under dry
conditions. Therefore, the design requirement for ready retrieval in 10
CFR 72.122(l) does not provide a basis for issuing an order as
requested by the Petitioner.
Similarly, requiring the licensee to maintain space in the spent
fuel pool is not necessary as a contingency for certain events or
conditions for which a cask must be returned to the spent fuel pool to
facilitate inspections or ensure adequate cooling of the fuel
assemblies. During its reviews performed during certification of the
VSC-24 design, the NRC staff confirmed that the design features of the
cask provide reasonable
[[Page 19822]]
assurance that the cask and fuel assemblies will confine the
radioactive materials following the design basis events established for
dry storage casks. These design features include the confinement
function provided by the welded MSB, the cooling and shielding
functions provided by the ventilated concrete cask (VCC), the
limitations on the fuel to be stored, and other cask characteristics
and limitations placed on its use that were relied upon during the
NRC's certification of the cask. Although the NRC staff considered it
prudent to require a cask to be returned to the spent fuel pool to
ensure cooling of the spent fuel and support inspections to confirm
that the cask could remain in service following certain design basis
events, the ability of the VSC-24 casks to withstand such events made
it unnecessary for the NRC to include specific time constraints in
which the operation needed to be completed.3
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\3\ The position that a time-urgent unloading of a cask need not
be considered is also supported by the analysis of a hypothetical
event involving the failure of the stored fuel pins with subsequent
ground level breach of an MSB that was presented in the SAR for the
VSC-24 design. Although no identified accident results in such
failures, the event was analyzed to demonstrate the limited
radiological consequences from accidents involving VSC-24 casks.
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In the event that a condition would arise requiring a cask to be
returned to the spent fuel pool, the continued confinement of the
radioactive materials within the MSB would afford the licensee ample
time to develop corrective actions that would maintain safe storage
conditions and minimize occupational exposures. The design features of
the cask, the unlikely nature of events that may require unloading a
cask, and the NRC staff's judgment that licensees could develop an
alternate approach if a spent fuel pool could not support an immediate
unloading of a cask have previously been cited as reasonable
justification for not requiring licensees to maintain a fixed reserve
capacity in spent fuel pools.4
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\4\ See resolution of public comments published with rulemakings
to add the VSC-24 cask (58 FR 17948) and TN-24 cask (58 FR 51762) to
the list of NRC-certified casks.
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Requirements defining conditions for returning a cask to the spent
fuel pool were included in the COC for the VSC-24 cask in order to
maintain the cask components and stored spent fuel assemblies within
the boundaries evaluated and accepted by the NRC staff during the
certification process. The COC addresses those events or conditions
which might lead to degradation of the cask or fuel assemblies. The
required actions normally include restoring operations to within the
acceptable limits or otherwise ensuring the spent fuel is placed in a
safe storage condition. The COC requirements for some events or
conditions include returning the MSB to the spent fuel pool to provide
a safe storage condition and unloading of the spent fuel assemblies in
order to support inspections of the cask.
The COC-required action in Section 1.2.10, ``Time Limit for
Draining the MSB,'' states that a cask should be returned to the spent
fuel pool for cooling if the water cannot be drained within the
specified time after the MSB is removed from the spent fuel pool with
24 spent fuel assemblies. The referenced draining operation is part of
the cask-loading sequence and it is reasonable to assume, therefore,
that the cask-loading area within or adjacent to the spent fuel pool
would be available for the cask should this contingency need to be
implemented. Further, the COC-required action is meant to restore
cooling to maintain safety margins pertaining to fuel assembly
subcriticality and can be accomplished without unloading the fuel
assemblies from the MSB. It is likely, however, that the locations in
the spent fuel pool that had contained the fuel assemblies loaded into
the storage cask would remain available during the loading and draining
of the cask.
Section 1.2.15, ``Handling Height,'' requires fuel assemblies to be
returned to the spent fuel pool, and inspections and evaluations
performed for cask components in the event a loaded cask is dropped
from a height greater than 18 inches. The COC prohibits handling of a
loaded VCC at a height greater than 80 inches. The NRC evaluation of
the MSB drop analysis concurred that drops up to 80 inches of the MSB
inside the VCC can be sustained without breaching the confinement
boundary, preventing removal of the spent fuel assemblies, or causing a
criticality accident. However, it is deemed prudent to return the cask
to the spent fuel pool to perform inspections and evaluations in the
event a cask experiences a significant drop, which is considered to be
a drop from a height greater than 18 inches. The requirement to perform
such inspections and evaluations was, therefore, included in the COC in
the event that a cask were to be dropped during movement. However,
since the most likely time for a cask drop event to occur would be
during movement of a newly loaded cask to the ISFSI, it is reasonable
to assume that the spaces in the spent fuel pool that had contained the
fuel assemblies loaded into the cask would remain available. Moreover,
even assuming for the sake of this analysis that the drop occurs when
spaces might not be available in the spent fuel pool, reviews of the
cask have shown that the cask and fuel will remain intact following a
drop from the maximum allowable height. Because a drop from the maximum
allowable height would not pose an immediate threat to the safety of
the public or plant personnel, adequate time would be available for the
licensee to develop and implement approaches to perform the required
inspections and evaluations if spaces were not available in the spent
fuel pool to support an immediate unloading of the cask. Temporary
shielding, loading the affected MSB into a spare VCC, placing the
affected MSB into the cask loading area within or adjacent to the spent
fuel pool, or other contingency actions could ensure safe storage
conditions while the licensee developed and implemented an approach to
allow for the actual unloading of the cask that had been dropped.
The requirements contained in Sections 1.2.3, ``Maximum Permissible
Air Outlet Temperature,'' and 1.3.4, ``Thermal Performance,'' were
included in the COC to provide reasonable assurance that the
temperatures of the fuel cladding and the VSC-24 concrete do not exceed
design limits. Concrete temperature limits are intended to prevent
gradual degradation of the VCC and the shielding it provides for the
MSB, which is the containment vessel for the spent fuel. Other
temperature limits pertain to the fuel cladding and are intended to
maintain the stored fuel assemblies below the temperatures at which
damage might occur. However, in the event that excessive temperatures
are detected, cooling of the cask and subsequent placement of the MSB
into the spent fuel pool, if necessary, are sufficient to avoid
immediate safety concerns. Because safe storage of the fuel assemblies
is achieved by placing the affected MSB into the cask loading area
adjacent to or within the spent fuel pool, the actual unloading of the
assemblies from the MSB to the storage racks within the spent fuel pool
can await the licensee's development of alternative approaches if that
were necessary due to a lack of storage space in the spent fuel pool.
Such approaches may require the licensee to make modifications to the
spent fuel pool or other parts of the reactor facility.
In addition to the specific COC requirements previously discussed,
a cask might need to be returned to the spent fuel pool if the cask
fails to meet some criteria provided in NRC regulations or the COC and
should, therefore, be removed from service.
[[Page 19823]]
Tests and surveillances performed before and after loading spent fuel
into a storage cask are designed to detect failures to conform to
design or regulatory requirements before a problem presents an imminent
threat to the cask or stored fuel. Therefore, while discovery of a
nonconformance or previously unidentified vulnerability may require
removing a cask from service as part of a licensee's corrective
actions, it is highly improbable that the discovery of such a condition
would pose an immediate safety concern. As in the previous examples,
safe storage of the spent fuel could be accomplished by returning the
affected MSB to the cask loading area within or adjacent to the spent
fuel pool and the MSB and spent fuel could remain there while the
licensee determined an appropriate course of action, including
provisions for unloading the cask, if necessary.
In sum, no credible accident has been identified that would require
the immediate unloading of a storage cask as a necessary protective
measure to avoid significant radiological consequences to members of
the public. In addition, there is no event or condition that was
identified during the certification of the VSC-24 cask that would
require a time-urgent unloading of a cask. Therefore, there is no need
for NRC to require continuous availability of space in the spent fuel
pool to accommodate the potential need to unload a cask. Further, the
NRC staff has reasonable assurance that licensees could, if necessary,
develop and implement an approach to unload a cask if required to do so
by unplanned events or conditions, such as those identified in the COC.
If space is not immediately available in the spent fuel pool, there
would be time to make it available by relocating other spent fuel
assemblies or removing them for temporary storage in a cask or by
making modifications to the spent fuel pool or other parts of the
reactor facility. Therefore, the NRC does not see a need to require the
licensee to reserve a fixed number of vacant spaces in the spent fuel
pool or to maintain the capability to retrieve the spent fuel from a
cask within a specified period of time, particularly when there is no
such prescriptive requirement stated in NRC rules.
Item 2: Order VSC-24 Users Not To Load Casks Pending Amendment of
Documents
The Petitioner's second request was for the NRC to order all users
of the VSC-24 cask not to load VSC-24 casks until the COC, the SAR, and
the SER are amended to contain operating controls and limits that
prevent hazardous conditions. As noted previously, following the event
at Point Beach, the NRC staff recognized that additional evaluation of
potential material interactions was warranted for all transportation
and storage casks. In regard to the VSC-24 cask, the event and
subsequent NRC inspections made it apparent that actual changes in the
operating procedures or the design of the cask would be necessary. CALs
were issued to confirm licensees' commitments to refrain from loading
VSC-24 casks pending completion of the staff's review of the responses
to NRC Bulletin 96-04 and verification of the associated corrective
actions. As discussed, the CALs established a process by which the NRC
staff could obtain confidence that operating controls and limits to
address potential hazardous conditions are developed and implemented by
each licensee using VSC-24 casks.
In particular, the CAL process ensures that licensees will
incorporate the necessary operating controls and limits into revised
plant procedures. Moreover, under existing NRC requirements, the
licensee must adequately implement those revised procedures. For this
reason, no changes to the COC or the SAR are needed to ensure that
enforceable operating controls and limits are in place to address
potential hazardous conditions during the loading or unloading of a
cask. Further, as previously indicated, the staff has documented the
process, information, and results of its review of the licensee's
response to Bulletin 96-04 for use of the VSC-24 at ANO and Point Beach
in safety evaluations available for public review. The NRC staff is
currently reviewing the responses to the bulletin submitted by the
licensee for Palisades.
Although the actions taken as part of the CAL process provide
adequate assurance that technical and regulatory compliance issues
raised by the event at Point Beach will be resolved before a licensee
loads or unloads a VSC-24 cask, the NRC staff agrees with the
Petitioner that it would be beneficial if the SAR and other licensing
basis documents accurately described the identified chemical reaction
and the associated operating controls and limits. The NRC staff is
currently reviewing a proposed amendment to the SAR and the COC for the
VSC-24 cask design and will ensure that the information related to the
identified chemical reaction and associated operating controls is
adequately addressed in the appropriate licensing-basis document. In
addition, the NRC staff is processing a petition for rulemaking, PRM-
72-3, that may lead to additional updating of ISFSI SARs and the
inclusion of information on operating controls and limits implemented
as a result of the event at Point Beach. However, the previously
discussed controls to be implemented by the licensees and verified by
the NRC staff as part of the CAL process, and the enforceability of
those controls under existing NRC requirements, make it unnecessary to
require revision of the specific licensing documents cited by the
Petitioner as a precondition for resuming cask operations at the
facilities using VSC-24 casks.
IV. Conclusion
The Petitioner requested that the NRC (1) Require WEPCO to retain
24 empty and available spaces in the Point Beach Nuclear Plant spent
fuel pool to accommodate retrieval of spent fuel from a VSC-24 cask,
and (2) prohibit loading of VSC-24 casks until the COC, the SAR, and
the SER are amended to contain operating controls and limits to prevent
hazardous conditions. Each of the claims by the Petitioner has been
reviewed. I conclude that for the reasons discussed above, no adequate
basis exists for granting the Petitioner's request for either (1)
Requiring the licensee for Point Beach to reserve a fixed number of
vacant spaces in the spent fuel pool or (2) suspension of the
licensees' use of the general license for dry cask storage of spent
nuclear fuel at Palisades, Point Beach, or Arkansas Nuclear One pending
revision of the SAR, the SER, and the COC for the VSC-24 cask.
A copy of this decision will be filed with the Secretary of the
Commission for the Commission to review in accordance with 10 CFR
2.206(c). As provided by this regulation, this decision will constitute
the final action of the Commission 25 days after issuance unless the
Commission, on its own motion, institutes a review of the decision
within that time.
Dated at Rockville, Maryland, this 17th day of April 1997.
For the Nuclear Regulatory Commission.
Samuel J. Collins,
Director, Office of Nuclear Reactor Regulation.
[FR Doc. 97-10522 Filed 4-22-97; 8:45 am]
BILLING CODE 7590-01-P
